Abschlussbericht Reaktorsicherheitsforschung - Vorhaben-Nr.: 150 1331 Wissenschaftlich-technische Zusammenarbeit mit Russland: Untersuchungen zu Mechanismen der Strahlenversprödung und des Ausheilverfahrens sowie Fluenzberechnungen für Reaktordruckbehälter von WWER-Reaktoren

The project was performed in the framework of the scientific technical cooperation in the scope of nuclear safety research between Germany and Russia. Objects of the investiga-tions are the decommissioned reactor pressure vessels (RPV) of the Greifswald nuclear power plant. The Greifswald WWER-440/V-230 nuclear reactors represent the first genera-tion of this reactor type. The investigation of these RPV’s enable the assessment of the aging and the effect of an industrial thermal annealing of serial RPVs for the first time. The main focus of the investigations was not on the application of mechanical-technological test methods on which the Russian technical regulation is based, but on the application of ad-vanced fracture mechanics test methods. This enables a fracture mechanics RPV integrity assessment which is based on direct measured fracture toughness values. With the investi-gations it is shown that real neutron induce embrittlement of the beltline welding seam and the forged base metal ring cannot be predicted by the Russian technical regulation. The direct measured KJC values and the resulting Master Curve T0 values, which characterise the ductile-to-brittle transition, differ fundamentally of those which were determined on mechani-cal technological values.
A further topic is the investigation of irradiation induced microstructural defects which origi-nates the embrittlement and their mitigation by the thermal annealing. Within separate sub-projects neutron fluence calculations were performed for the investigated WWER-440/V-230 RPVs and their support constructions.